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Licensing Issues and the PIRT

Frederik Reitsma

Oct 22-26, 2012

IAEA Course on High temperature Gas Cooled Reactor Technology

Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 2

Content / Overview

– A few ideas to stimulate discussions:

– Safety assessment criteria

– Safety analysis

– Treatment of uncertainties

– NRC Advanced Reactor Policy

– Evaluation models

– PIRT


Safety assessment • Combination of both deterministic and probabilistic methods

• South African National Nuclear Regulator has set the following limits:


Safety analysis • Combination of both best estimate and conservative deterministic models and

analysis – best estimate analysis, the best estimate material properties and plant parameters are used

as inputs to the analysis. Other sources of uncertainty may be addressed more conservatively

– Conservative analysis results are achieved using conservative inputs in conservative models. Sensitivity analyses are often used to ensure parameters are set to ensure pessimistic results with respect to the acceptance criterion.

• Best estimate analyses are used: – to demonstrate As Low As Reasonably Achievable (ALARA) for Anticipated Operational

Occurrences (AOOs) and Design Basis Accident (DBAs).

– to determine the expected consequences of more hypothetical accidents, i.e. Beyond Design Basis Accidents (BDBAs)

– to provide the Probabilistic Risk Assessment (PRA) with expected or realistic consequences instead of conservatively biased ones.

• Conservative analyses are used: – To demonstrate compliance with regulatory limits for AOOs and DBAs


Treatment of uncertainties • Combination of both best estimate and conservative deterministic models and analysis

– best estimate analysis, the best estimate material properties and plant parameters are used as inputs to the analysis. Other sources of uncertainty may be addressed more conservatively

– Conservative analysis results are achieved using conservative inputs in conservative models. Sensitivity analyses are often used to ensure parameters are set to ensure pessimistic results with respect to the acceptance criterion.

• Best estimate analyses are used: – to demonstrate As Low As Reasonably Achievable (ALARA) for Anticipated Operational Occurrences

(AOOs) and Design Basis Accident (DBAs).

– to determine the expected consequences of more hypothetical accidents, i.e. Beyond Design Basis Accidents (BDBAs)

– to provide the Probabilistic Risk Assessment (PRA) with expected or realistic consequences instead of conservatively biased ones.

• Conservative analyses are used: – To demonstrate compliance with regulatory limits for AOOs and DBAs

• Methodology based on the Code Scaling, Applicability, and Uncertainty (CSAU) process has been adopted to quantify uncertainties in best-estimate calculations – Can by used to show the margin of conservative models and analysis

– The GRS-SUSA code to be used as point of departure.

• A new IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis is being performed.


NRC Advanced Reactor Policy Statement (1/2)

“Among the attributes which could assist in establishing the acceptability or licensability of a proposed advanced reactor design, and which therefore should be considered in advanced designs, are:

• Highly reliable and less complex shutdown and decay heat removal systems. – The use of inherent or passive means to accomplish this objective is

encouraged (negative temperature coefficient, natural circulation).

• Longer time constants and sufficient instrumentation to allow for more diagnosis and management prior to reaching safety system challenge and/or exposure of vital equipment to adverse conditions.

• Simplified safety systems which, where possible, reduce required – operator actions, – equipment subjected to severe environmental conditions, – and components needed for maintaining safe shutdown conditions. – Such simplified systems should facilitate operator comprehension,

reliable system function, and more straight-forward engineering analysis for analysis.


NRC Advanced Reactor Policy Statement (2/2)

• Designs which minimize the potential for severe accidents and their consequences by providing sufficient inherent safety, reliability, redundancy, diversity, and independence in safety systems.

• Designs that provide reliable equipment in the balance of plant, (or safety-system independence from balance of plant) to reduce the number of challenges to safety systems.

• Designs that provide easily maintainable equipment and components. • Designs that reduce radiation exposure to plant personnel. • Designs that incorporate defense-in-depth philosophy by maintaining

multiple barriers against radiation release, and by reducing the potential for consequences of severe accidents.

• Design features that can be proven by citation of existing technology or which can be satisfactorily established by commitment to a suitable technology development program.”

FR Vol 73 No. 199, pg. 60612-60616, Oct. 14, 2008


Evaluation Models


PBMR Evaluation Models

• Pre-Break – Focuses on the expected pre-break conditions, just

before a break in the helium pressure boundary occurs.

– The time phase for this Evaluation Model ends when such a break occurs.

– It does not include any of the phenomena that might occur during a pressure boundary break or later.

– Software used: VSOP, MCNP, FLOWNEX, Fluent, NobleG, FIPREX/GETTER, RADAX.



• Reactivity transients

– Calculates the transient reactor response for reactivity transient scenarios.

– Number of calculation models: 2

– Software used: VSOP, TINTE.

• Thermal transients

– Calculates the transient reactor temperatures for forced cooling and loss of forced cooling scenarios.

– Number of calculation models: 2.

– Software used: TINTE, FLOWNEX.


PBMR Evaluation Models • Normal operation release

– Calculates the release of activity during normal operation as a result of Helium Pressure Boundary leakage.

– Number of calculation models: 22 – Software used: VSOP, MCNP, FLOWNEX, ASTEC, Fluent,

NobleG, FIPREX/GETTER, RADAX. • Maintenance Dose

– Calculates the dose received by maintenance workers during maintenance periods due to the dust in the MPS and the activation of components that took place during normal operation.

– Number of calculation models: 15 – Software used: MCNP, RADAX, SCALE, FISPACT,

MicroShield.



• Overall Worker and Public Dose – Calculates the doses received by the public and worker for

Category A, B and Beyond B accidents that involve a breach in the pressure boundary.

– Provides the source term for each of these accidents as well as the resulting doses.

– Provides source terms that are used as an input into the Probabilistic Risk Assessment (PRA) for Category C events.

– Number of calculation models: 28

– Software used: VSOP, TINTE, MCNP, ASTEC, FLOWNEX, Fluent, PC COSYMA, NobleG, FIPREX/GETTER, RADAX.

– Main components: Pre-break, Initial release, Delayed release, Air ingress, Confinement, Atmospheric dispersion


PIRT

Phenomena Identification Ranking Tables


PIRT background • Phenomena Identification and Ranking Tables are increasingly used in the nuclear

industry. • PIRT were initially used to identify the thermal-hydraulic processes that were most

important for a safety analysis computer code to simulate with acceptable accuracy, so that a limited set of sensitivity analysis could be performed to help quantify the uncertainty in the safety analysis results.

• PIRT are now recognized as a valuable tool to help prioritise efforts associated with safety analysis, development and assessment of codes and models, and specification of scaling or other requirements for tests and experiments.

• There is limited industry knowledge and experience with HTGR accident analysis, relative to that for LWRs.

• Since operating history is not available to provide the same valuable data, an accepted and auditable method of making early decisions related to analysis is needed.

• The PIRT process can be used as: – a tool and a guide to help prioritise software and model V&V efforts, – to agree on the necessary degree of conservatism to include in analysis assumptions and initial

conditions, – and to focus the available resources on the phenomena believed to be most important to the

safety analysis process.


PIRT - Process • Process:

– Step 1. Define the PIRT objectives and plant design – Step 2. Define the accident or transient scenario – Step 3. Define figures of merit – Step 4. PIRT team review available data – Step 5. Partition scenario into convenient time

phases – Step 6. Identify involved and affected SSC – Step 7. Identify phenomena by time phase and SSC – Step 8. Rank importance of components and

phenomena with confidence levels – Step 9. Finalize and document PIRT for subject

scenarios and plant designs

– Repeat periodically


PIRT Status Decision Chart Rank

Confidence

in Rank

Confidence

in Value

(High / Low)(Sure /

Unsure)

(Sure /

Unsure)

8 High Unsure Unsure

Phenomenon is perceived as

significant but is not well

known.

High priority requirement for

analysis and validation.

7 High Sure UnsurePhenomenon is significant

and confidence in value is low.


validation.

6 High Unsure Sure

Phenomenon is significant

and the confidence in rank is

low.


analysis.

5 High Sure SurePhenomenon is significant

and well known.

Should be well represented in

the model. Should be readily

validated.

4 Low Unsure UnsurePhenomenon is not significant

but not well known.

Requires analysis and

validation to determine rank

and value.

3 Low Sure Unsure

Phenomenon is not significant

and the confidence in value is

low.

Low priority requirement for

validation.

2 Low Unsure Sure

Phenomenon is not significant

and the confidence in rank is

low.

Low priority requirement for

analysis.

1 Low Sure SurePhenomenon is well known

and is not significant.

May be modelled without

validation.

Status Symptom Action required


PIRT - examples

• Source Term (Pre-Break)

• Reactivity transients

• Thermal transients

• Overall Worker and Public Dose PIRTs – Initial release

– Delayed release

– Air Ingress

– Confinement

– Atmospheric Dispersion

• Non-MPS leaks and spills


PIRT example Transport phenomena (from pre-break PIRT)

• Radionuclide plateout: Adsorption/ desorption processes Penetration/evaporation Diffusion into material Chemical characteristics of radionuclide Laminar or turbulent flow

• Dust deposition and lift-off: Agglomeration of dust particles Brownian diffusion Electrostatic forces Inertial separation Laminar or turbulent flow Saffman lift force Sedimentation Thermal gradient (thermophoresis) Plant operational transients Plant vibration Monolayer or multilayer resuspension

Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology

Reactivity Transient PIRT Example for the Reactor Unit

Ref

Number

System subsystem component Process Phenomena Rank

(3 - high,

2 -

medium,

1 - low)

Confidenc

e (Sure/

Unsure)

state of

knowledg

e

Rationale Notes

RT-RU-001 RU Pebble Bed Fuel Kernel Fission Fission Power

production in the fuel

High Sure High The fission process provides the

primary source of power production

in a reactor.

This is due to fission,

RT-RU-002 RU Pebble Bed Fuel Kernel Fission Temperature

influence on fission

(doppler effect)

High Sure High Kernel does get hot. Flux

dependence on group structure is

important. Temperature

dependence of resonance cross

sections important.

Consider Multi-group vs 2-group

neutron flux representation. The current

2 group misses out on resonance

treatment. The physics is understood,

but it comes down to modelling issues

and complexity.

RT-RU-003 RU Pebble Bed Fuel Pebble Fission Moderation -

temperature

dependence

High Sure High Moderator feedback effect is well

known and temperature interaction

with fission is well modelled

The moderator feedback effect is

known and temp interaction with fission

is currently well modelled.

RT-RU-004 RU Pebble Bed Coolant Fission Nitrogen inventory

reduction

High Sure Med Nitrogen is an absorber. The

reduction in the nitrogen inventory

could contribute to reactivity

addition.

This is particularly applicable to the

start-up sequence, the the reduction in

nitrogen could contribute to reduced

absortion. The effect of SAS removal

with nitrogen in the core needs to be

checked to see if the core approaches

criticality. The modelling of the scenario

has never been attempted but physics

capability is available in TINTE.

RT-RU-005 RU Pebble Bed Fuel Kernel Fission Non-local Power

Production

Med Sure High The negative coefficient of

reactivity dependence on

temperature is assisted by non-

local heating which makes this

phenomenon important.

The non-local power is as a result of

absorption of gamma radiation, fast

neutrons. Currently when modelling,

the fission process is adequately

captured, but gamma modelling is

approximated.

RT-RU-006 RU Pebble Bed Fuel Kernel Fission Fuel Burn-up Low Sure High Low for perturbations in burn up

from the anticipated core state.

Refers to the average value of Fuel

Burn-up (there will be differences when

considering the start-up or equilibrium

core)

RT-RU-007 RU Pebble Bed Fuel Kernel Fission Spatial distribution of

burn-up

Low Sure High Low for perturbations in burn up

from the anticipated core state.

Currently the axial distribution within

the core is modelled.

RT-RU-008 RU Pebble Bed Reactor core

incl graphite

SSC

Fission Nuclei Breeding Low Sure High The total effect on the FOM is

minimal

Nuclei Breeding is captured in the

characterisation of power production.

The effect is predominantly considered

over the whole core and not to the

kernel level. 19


PIRT – Iterative process • Positive outcomes:

– Good basis for identifying EM development requirements

– Easier to justify modelling assumptions • Challenges:

– In a new technology experts are not readily available • Spent a lot of time on irrelevant phenomena • Ranking may be incorrect • If nobody knows then typically have to spend a lot of time to

find out … but in the end this is positive – Include external experts

• Improves credibility (but also complexity…) – Perform hierarchical breakdown – Revised ranking bins

• Difficult to decide what to do with “medium” bins • Too many combinations of uncertainties available to

adequately action resolution

Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology

• Source material used: – HTGR Technology Course for the Nuclear Regulatory Commission,

May 24 – 27, 2010

– HTR/ECS 2002 High temperature Reactor School, 2002

– MUA 784: Reactor Physics, F Reitsma, Mechanical Engineering Post-Graduate: Nuclear Theme, University of Pretoria, 2012

– Workshop at PHYSOR 2010 – Advances in Reactor Physics to Power the Nuclear Renaissance: The Pebble Bed Modular Reactor: From V.S.O.P. (Very Superior Old Product) to Generation–IV candidate.

– “Safety Analysis Software Development and V&V”, Peter Robinson, Workshop on Safety Aspects of Modular HTGRs, October 2007, Beijing China

– Radionuclide transport during normal operation conditions”, Lize Stassen, Pieter Goede, Gen-IV CMVB Chemistry and Transport Workshop, Centurion, South Africa, 12 – 13 January 2009

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